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Core materials development for the fuel cycle R&D program
S. A. Maloy
, M. Toloczko
, J. Cole
, T. S. Byun
Nuclear Fuels & Materials
Research output
:
Contribution to journal
›
Article
›
peer-review
38
Scopus citations
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Keyphrases
Fuel Cycle
100%
Material Development
100%
R&D Programs
100%
Coolant
100%
High Temperature
100%
Irradiation
100%
HT-9
100%
Oxide Dispersion
100%
Reactor
50%
High Burnup
50%
Yield Stress
50%
Development Program
50%
Minor Actinides
50%
Fast Reactor
50%
Fuel Burnup
50%
Burnup
50%
Temperature Range
50%
Mössbauer Spectroscopy
50%
Mechanical Testing
50%
Radiation Effects
50%
Ferritic-martensitic Steel
50%
Radiation Resistance
50%
High Dose
50%
Transmutation Fuel
50%
Radiation Tolerance
50%
Radiation-induced Segregation
50%
Chemical Interaction
50%
Total Fluence
50%
Compact Tension Specimen
50%
Total Dose
50%
Fast Reactor Core
50%
Compression Testing
50%
Low-temperature Embrittlement
50%
Creep Corrosion
50%
Helium Embrittlement
50%
Charpy Testing
50%
Accelerated Creep
50%
C-compact
50%
Compact Tensile Specimen
50%
Charpy
50%
Engineering
Burnup
100%
Fuel Cycle
100%
Material Development
100%
Duct
100%
Core Material
100%
Fast Reactor
66%
Embrittlement
66%
Charpy
66%
Yield Point
33%
Low-Temperature
33%
Minor Actinide
33%
Reactor Core
33%
Temperature Range
33%
Creep
33%
Radiation Effect
33%
Mechanical Testing
33%
Test Specimen
33%
Tensile Specimen
33%
Compression Testing
33%
Moessbauer Spectroscopy
33%
Mossbauer Spectroscopy
33%
Material Science
Oxide Compound
100%
Coolant
100%
Embrittlement
100%
Corrosion
50%
Yield Stress
50%
Actinide
50%
Creep
50%
Mechanical Testing
50%
Mossbauer Spectroscopy
50%
Compression Testing
50%